Nuclear reactors have their fuel contained in sealed cladding for the isolation of the nuclear fuel from the moderator/coolant system. The term cladding, as used herein, refers to a zirconium based alloy tube composed of at least one metal in addition to the zirconium base. The term precipitates, as used herein, refers to added metals of the cladding and forming isolated structures in a matrix throughout the zirconium alloy. These precipitates may or may not constitute intermetallics. Typically, these precipitates are uniformly distributed in the matrix--although they vary in size. Further, so-called fine precipitates (below 0.1 microns), can either be in the matrix format or the so-called two dimensional format where the precipitates occupy sheet like layer near the outer surface of the zirconium alloy.
The cladding--nominally in the order of 0.030 inches thick--is formed in the shape of a tube with the nuclear fuel contained typically in pellet form therein. These pellets are stacked in contact with one another for almost the entire length of each cladding tube, which cladding tube is in the order of 160 inches in length. Typically, the cladding tube is provided with springs for centering the fuel pellets and so-called "getters" for absorbing fission gases. Thereafter, the internal portions of the fuel rod are pressurized with various gases for optimum dissipation of gases produced from the fission reaction, and sealed at both ends.
Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel cladding since they have low neutron absorption cross sections and at temperatures below about 398.degree. C. (at or below the core temperature of the operating reactor) are strong, ductile, extremely stable and nonreactive in the presence of demineralized water or steam. "Zircaloys" are a widely used family of corrosion-resistant zirconium alloy cladding materials. The Zircaloys are composed of 98-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are two widely-used zirconium-based alloys for cladding. (Zircaloy-4 omitting nickel).
Cladding corrosion is a potential problem both in boiling water reactors and pressurized water reactors. For example, in a PWR, water does not boil--although in modern designs minute boil can occur at the top of some fuel rods. The oxygen level is relatively suppressed, being about 20 ppb. Hydrogen is injected and resident in the water moderator at about 200 ppb and utilized to suppress oxygen levels. Water pressure is in the range of 2000 psi with temperature ranging from 300.degree. C. to 380.degree. C. dependant upon the operating state of the reactor.
Corrosion in PWR cladding is uniform and related to precipitate size in the Zircaloy cladding. Small precipitates have been found to actually accelerate the uniform corrosion phenomena. Consequently, relative large precipitate sizes are preferred in the PWR zirconium cladding.
In the radiation environment within the PWR, the precipitates dissolve and become smaller with exposure. To avoid accelerated uniform corrosion buildup, PWR cladding uniformly starts with large precipitate sizes--0.2 microns and above--to slow the formation of small size precipitates and the more rapid uniform corrosion that occurs with the small size precipitates.
In a BWR environment, water does boil. The oxygen level is relatively high, being about 200 ppb. Hydrogen may be injected for the stability of structural parts of the reactor, is effectively stripped off as a part of the boiling, and is resident in the water moderator in the range of 20 ppb. Water pressure is in the range of 1000 psi with temperature at 288.degree. C. being essentially a function of pressure and for the most part constant all operating rates of the reactor.
Corrosion in a BWR occurs in nodular or pustule formats on the zirconium cladding. Uniform corrosion is also present--but in the usual case not to a significant degree. Further, mineral and particle deposition occurs on the water exposed surface of the cladding. The combination of the corrosion and depositions can become fairly thick on the water exposed portions of the cladding.
Nodular or pustule corrosion is not inherently bad. However, where fuel in the reactor has longer life--such as time within the reactor exceeding 40 megawatt days per ton, nodular or pustule corrosion concentrates. Where such nodular or pustule corrosion becomes concentrated and acts in conjunction with other particles--such as copper ions--localized penetration of the cladding wall can occur.
Small precipitates have been found to actually suppress nodule and pustule formation. Consequently, it is desired to have small precipitates--below 0.1 microns--to inhibit formation of nodules or pustules. It is known in the prior art to externally treat the outer water exposed surface of cladding with heating from a coil to produce a fine precipitate exterior surface. See Eddens et al. U.S. Pat. No. 4,576,654.
In the radiation environment within the BWR, the precipitates dissolve and become smaller with radiation exposure. Nodular corrosion is inhibited by the small precipitates and by the alloying elements put in solution by the dissolution process.
Anneals of zirconium alloys have been used and can be summarized in terms of temperature ranges. Starting at low temperatures, anneals above 480.degree. C. effect stress relief, usually after working of the metal to achieve around 70% reduction in area. Anneals at about 576.degree. C. not only effect stress relief but also commence recrystallization of the metal. In such anneals, maximum ductility of the material is achieved. Finally, anneals substantially above 576.degree. C. effect crystal growth--generally softening the metal.
In the prior art, the heat treatment for PWR cladding has included high temperature anneals with slow quenches (less than 5.degree./second) to preserve large precipitate sizes. Conversely, the heat treatment for BWR cladding has included low temperature anneals with fast quenches (greater than 5.degree./second) to produce small precipitate sizes.
The corrosion resistance of Zircaloy cladding has been improved by forming small, uniformly distributed precipitate particles in the Zircaloy metal matrix. Some portion of the iron, chrome, and nickel components in the Zircaloy matrix form insoluble crystalline precipitates having chemical compositions distinct from the matrix. The precipitates are generally represented by the chemical formulas Zr(Fe,Cr).sub.2 and Zr.sub.2 (Fe,Ni). Typically the precipitates used in the more corrosion resistant alloys have an average diameter of less than about 0.1 microns.
Corrosion and cracking can both damage cladding, but they are fundamentally different phenomena. Cracking is a mechanical breaking or splitting of the cladding wall, while corrosion is an electrochemical conversion of the cladding metal into an oxide or other non-metallic compound. Cracks may be initiated by a variety of causes including mechanical stresses as well as corrosion. Once a crack is initiated, it may pose little problem, so long as it remains confined to a small area. However, if the crack propagates, the cladding can be breached and the fission material eventually contacts the coolant or moderator. Ultimately, this can lead to an expensive reactor outage.
The mechanical initiation of cracks can be attributed to various stresses in a conventional reactor. Cracks can start when debris such as wires or metallic shavings or particles find their way into reactor water that flows within the fuel bundles between the fuel rods. The debris may lodge at a fuel rod spacer adjacent the cladding wall. As a result, the debris vibrates or frets against the cladding wall under the influence of the passing steam/water mixture. Such vibration continues until a crack begins.
Corrosion can be the source of initial crack propagation. Moreover, manufacturing defects can be the points of crack origin. Still further, crack propagation can start on the inside of the fuel rods in the corrosive high pressure environment present during in service reactor life.
Regarding cracking in the interior of the sealed cladding tube, brittle splitting of such cladding may occur due to the combined interactions between the nuclear fuel, the cladding, and the fission products produced during the nuclear reaction. It has been found that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. These localized stresses and strain in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking and liquid metal embrittlement. Other phenomena such as local hydriding of the cladding and the presence of oxygen, nitrogen, carbon monoxide, and carbon dioxide can assist cladding failure and lead to rod cracking.
U.S. Pat. Nos. 4,200,492 and 4,372,817 to Armijo et al as well as Adamson U.S. Pat. No. 4,894,203 suggest solutions to preventing crack initiation by including a barrier on the inside of the cladding. Cladding containing introduce barrier are sometimes referred to as "composite" cladding or cladding having two distinct metallurgical layers.
Although it is highly desirable to prevent crack initiation, in the event a crack forms, its propagation is to be avoided.
There exists a need, especially for a BWR environment, for cladding which is resistant to axial crack propagation. There also exists a need for cladding which, in combination, is resistant to axial crack propagation, crack initiation and corrosion.